characterization of tritium transport in the flibe-graphite system€¦ · period ending july 31,...

1 Huali Wu ǁ‖ HEATandMASS.ep.wisc.edu

Characterization of Tritium Transport in the FLiBe-Graphite System

Huali Wu Michael Young, Nisarg Patel, Jayeesh Bakshi, Raluca Scarlat

Nuclear Engineering, UW [email protected] ǁ HEATandMASS.ep.wisc.edu

09.02.2015

Motivation

2 Heat & Mass Transport Group

!  Overview of FHR Technology

!  Tritium Inventory in Molten Salt Reactor and FHRo  Tritium Production o  Tritium in Graphite o  Tritium in Molten Salt

!  Graphite as a potential tritium sink in FHRo  MSRE experience o  FHR uniquenesso  Irradiation Effect o  Salt infiltration

!  Summary

!  Current/Future Work

Overview of FHR technology


UNIQUENESSLarge Carbon Surface Areao  Fuel pebbles and inert graphite

pebbles – 1945 m2 o  Inner and outer graphite reflector – 54

m2 o  Carbon filter cartridges (if needed) in

CTAH – 2300 m2 Nuclear Air-Brayton Combined Circle

Thermal Power 236 MWth Core Sphere Composition

68.3% Fuel Core Inlet Temperature 600 °C 31.7% Graphite

Core Outlet Temperature 700 °C Residence Time of Pebbles 2.1 Month Fuel Pebble Quantity 440000 Pebble Circulation Rate 10800 Pebbles/day

Inert Graphite Pebble Quantity 204000

CONCERNSo  Large tritium production rateo  Salt chemical control affects tritium

species (T2, TF, HT, etc.)o  Tritium Leakage through CTAH

and other metal structures

Defueling Machine

Hot Salt Extraction

DHX Wells

Central Reflector

Control Rods

Graphite Pebbles

Fuel Pebbles

Outer Reflector

Removable Vessel Head Heated Air Outlet Duct

Coiled Tube Bundle

Miter Bend Turning Vans

CTAH Pressure Vessel with Internal Insulation

Hot Salt Manifold Pipe

Cooled Salt Manifold Pipe Warm Air Inlet Duct

Tritium Inventory


TRITIUM PRODUCTION !  Tritium Source

o  6Li, 7Li, 9Beo  Initial Li-6 enrichment 60ppm → Tritium production rate 0.11mol/EFPDo  Equilibrium Li-6 concentration 4 ppm → Tritium production rate 0.023mol/EFPD

!  Tritium Emission Limit o  0.1pCi/L for air effluent, and 1µCi/L for water effluent

0.00

0.05

0.10

0.15

0.20

0

5

10

15

0 1 2 3 4 5 6 7 8 9 10 11 (m

ol T

/EFP

D)

Triti

um P

rodu

ctio

n R

ate

(Ci/

EFP

Y)

Effective Full Power Years

Estimated Tritum Production Rate in the Mk1 PB-FHR Core Type g/yr/Gwe Ci/yr/Gwe(*10^6)

PB-FHR 253 24.4

PWR 0.08 0.00078

CANDU 573 55.1

HTGR 2.63 0.252

PBMR 512 4.92

MSR 91.8 0.883

o  99.9% of tritium produced in FHR should be recovered

FHR DESIGN GOAL

Tritium Inventory


TRITIUM SOURCES AND SINKS

Tritium Inventory


TRITIUM IN GRAPHITE!  Pore diffusion

o  Molecule diffusiono  Physisorption and chemisorptiono  Low energy

!  Bulk diffusiono  Atom diffusion o  Chemisorption dominatedo  Depending on trapping sites energy o  Trapping-detrapping process

o  Dpore = 1.28E-11 @ 650 °C & Dbulk = 3.7E-21 @ 750 °C

o  Trap site 1 is hard for tritium to reach @ 700 °C

IN FHR

Pore IG-110(Ultrasonic Cleaned)

A3(Polished and Ultrasonic Cleaned)

Pore

Tritium Inventory


TRITIUM IN MOLTEN SALT

!  Tritium is produced in the form of T+ , and exists in salt as TF, T2, HT, etc. depending on Flibe property

!  TF is a condensable, corrosive gas!  T2 or HF is ready to diffuse through metal structural material @ 600 °C to 700 ° C!  Tritium will leak from tubes in CTAH

o  Aluminum oxide coating can reduce tritium leakage o  Cold trapo  Inert gas sparging

WHY SO IMPORTANT?

!  Via mass convection in Flibe, tritium will eventually reach surface of fuel pebbles and get absorbed or diffuse in graphiteo  DT2 ~ 3E-09 m2/s @ 650 ° Co  Saturation vs. rate-limiting

Graphite as A Potential Tritium Sink


SUMMARY OF MSRE EXPERIENCE

o  The only data set generated for Fluoride-salt-graphite system o  15% of tritium was absorbed by graphite and about half was within the first

1/16th inch layer

!  MSRE overviewo  8 MWth reactor with fuel salto  Operated at 700 °C o  The reactor core is made of nuclear graphite block

H. G. MacPherson, ‘Molten-Salt Reactor Program Quarterly Progress Report, for period ending July 31, 1960’, Oak Ridge National Lab, ORNL-3014, (1960)

FHR UNIQUENESS

!  Fluoride salt without fuel !  Higher power, higher neutron flux !  Larger carbon surface area!  Matrix graphite comprises ~40% in fuel pebble

APPLICATION OF MSRE DATA Fuel pebble



NUCLEAR GRAPHITE AND MATRIX GRAPHITE

Graphitization Process (http://www.graphiteconcept.com/content/view/33/27/)

o  Raw materialo  Heat treatment temperature

o  Microstructure o  Degree of graphitization o  Porosity o  Grain sizeo  Etc.


MATRIX GRAPHITE NUCLEAR GRAPHITE




IRRADIATION EFFECT

!  Increase the concentration of Trap 2 and Trap 1o  Irradiation has large effect on Trap 2

than Trap1 at low neutron fluence o  Trap 2 is more useful for tritium

retention in graphite due to its low activation energy

!  Irradiation effect on diffusion coefficient highly depends on neutron fluence

H. Atsumi, T. Tanabe, et al., ‘Hydrogen behavior in carbon and graphite before and after neutron irradiation – Trapping, diffusion and the simulation of bulk retention–’, J. Nucl. Mater., vol. 417, no. 1–3, pp. 633–636, (2011)

H. Atsumi, A. Muhaimin, et al., ‘Hydrogen trapping in neutron-irradiated graphite’, J. Nucl. Mater., vol. 386–388, pp. 379–382, (2009)



SALT INFILTRATION

o  Salt will diffuse into graphite, no dissolved carbon in salt observed from MSRE experiment

o  Impregnated salt in graphite tends to block or occupy pores or tritium diffuse paths

R. B. Briggs, ‘Molten-Salt Reactor Program Semianual Progress Report For Period Ending July 31, 1964’, Oak Ridge National Lab, ORNL-3708, (1964)

Salt

Summary


!  Tritium is produced in FLiBe by neutron irradiation of 6Li, 7Li and 6Be!  Large carbon surface area and continuously refueling in FHR could provide a perfect sink for

tritium!  Graphite property (graphitization, porosity, grain size, etc) will affect tritium retention !  Neutron irradiation has a positive effect on tritium retention in graphite due to the increase of

tapping sites!  Salt tends to penetrate into graphite and could play a negative role for tritium recovery with

graphite

SUMMARY


Current/Future Work

CURRENT/FUTURE WORK

2. Contact Angle measurement

1. Matrix Graphite Characterization

Graphite Sample Plate

Silica Glass shield with Insulation

Argon

Top surface without Insulation

o  Virgin graphite characterizationo  Post-irradiation measurement


3. Flibe-Intrusion Experiment

Vertical Furnace

Graphite Crucible

Sample Holder

Pressure Release Hole

4. Hydrogen behavior in Flibe-Graphite System

o  Hydrogen & graphite o  Hydrogen & Flibe-graphite system

Crucible lid


Questions?

characterization of tritium transport in the flibe-graphite system€¦ · period ending july 31,...

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